Effect of reprocessing on neutrons of a molten chloride salt fast reactor

Due to their unique features,such as the inherent safety,simplified fuel cycle,and continuous on-line reprocessing,molten salt reactors(MSRs)are regarded as one of the six reference reactors in the Generation IV International Forum(GEN-IV).Molten chloride salt fast reactors(MCFRs)are a type of MSR.Compared to molten fluoride salt reactors(MFSRs),MCFRs have a higher solubility of heavy metal atoms,a harder neutron spectrum,lower accumulation of fission products(FPs),and better breeding and transmutation performance.Thus,MCFRs have been recognized as a type of MSR with great prospects for future development.However,as the most important feature for MSRs,the effect of different reprocessing modes on MCFRs must be researched in depth.As such,this study investigated the effect of different isotopes,especially FPs,on the neutronic performance of an MCFR,such as its breeding performance.Furthermore,the characteristics of the different reprocessing modes and MCFR rates were analyzed in terms of safety,radioactivity level,neutron economy,and breeding capacity.In the end,a reprocessing method suitable for MCFRs was determined through calculation and analysis,which provides a reference for the further research of MCFRs.

Neutronic study on the effect of first wall material thickness on tritium production and material damage in a fusion reactor

In this study,the effects of changing first wall materials and their thicknesses on a reactor were investigated to determine the displacement per atom(DPA)and gas production(helium and hydrogen)in the first wall,as well as the tritium breeding ratio(TBR)in the coolant and tritium breeding zones.Therefore,the modeling of the magnetic fusion reactor was determined based on the blanket parameters of the International Thermonuclear Experimental Reactor(ITER).Stainless steel(SS 316 LNIG),Oxide Dispersion Strengthened Steel alloy(PM2000 ODS),and China low-activation martensitic steel(CLAM)were used as the first wall(FW)materials.Fluoride family molten salt materials(FLiBe,FLiNaBe,FLiPb)and lithium oxide(LiO_(2))were considered the coolant and tritium production material in the blanket,respectively.Neutron transport calculations were performed using the wellknown 3D code MCNP5 using the continuous-energy Monte Carlo method.The built-in continuous energy nuclear and atomic data libraries along with the Evaluated Nuclear Data file(ENDF)system(ENDF/B-V and ENDF/B-VI)were used.Additionally,the activity cross-section data library CLAW-IV was used to evaluate both the DPA values and gas production of the first wall(FW)materials.An interface computer program written in the FORTRAN 90 language to evaluate the MCNP5 outputs was developed for the fusion reactor blanket.The results indicated that the best TBR value was obtained for the use of the FLiPb coolant,whereas depending on the thickness,the first wall replacement period in terms of radiation damage to all materials was between 6 and 11 years.

Estimation of TBR on the Gap Between Neighboring Blanket Modules in the DEMO Reactor

When one wants to simply estimate tritium breeding ratio （TBR）, the TBR may be reduced from a ＂local＂ TBR for the breeding zones of a blanket module by multiplying the breeder coverage （= the surface area of effective breeding region / the surface area of the first wall around plasma）. When blanket modules are arranged~ the gap between neighboring modules and the frames of the modules are regarded as nombreeding zones. On the other hand, neutrons scattered in the non-breeding zones can enter breeding zones, contributing to tritium production. This means that the estimation method mentioned above tends to underestimate TBR. In order to assess the scattering effect quantitatively, we carried out a three-dimensional Monte Carlo N-particle transport MCNP-5 calculation. It was found from the calculation that there is little decrease in TBR for gaps less than 4 cm when the blanket thickness is 70 cm. The result indicates that such a wide allowance of the gap will facilitate access of remote handling equipment for the replacement of blanket modules and improve access of diagnostics.

Neutronic investigation and activation calculation for CFETR HCCB blankets

The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder（HCCB） blanket were predicted for the Chinese Fusion Engineering Testing Reactor（CFETR） design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio（TBR）. Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil.The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1？×？10-4 k W, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.